[Federal Register Volume 68, Number 198 (Tuesday, October 14, 2003)]
[Notices]
[Pages 59212-59229]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-25742]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission
(the Commission or NRC staff) is publishing this regular biweekly
notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of
1954, as amended (the Act), to require the Commission to publish notice
of any amendments issued, or proposed to be issued, under a new
provision of section 189 of the Act. This provision grants the
Commission the authority to issue and make immediately effective any
amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, September 19, 2003, through October 2,
2003. The last biweekly notice was published on September 30, 2003 (68
FR 56340).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission
[[Page 59213]]
take this action, it will publish in the Federal Register a notice of
issuance and provide for opportunity for a hearing after issuance. The
Commission expects that the need to take this action will occur very
infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By November 13, 2003, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714, which is
available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North,
[[Page 59214]]
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or
by e-mail to [email protected].
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: July 14, 2003.
Description of amendment request: The proposed change involves the
extension from 1 hour to 24 hours of the completion time (CT) for
Condition B of Technical Specification (TS) 3.5.1, which defines
requirements for accumulators. Accumulators are part of the emergency
core cooling system and consist of tanks partially filled with borated
water and pressurized with nitrogen gas. The contents of the tank are
discharged to the reactor coolant system if, as during a loss-of-
coolant accident, the coolant pressure decreases to below the
accumulator pressure. Condition B of TS 3.5.1 specifies a CT to restore
an accumulator to operable status when it has been declared inoperable
for a reason other than the boron concentration of the water in the
accumulator not being within the required range. This change was
proposed by the Westinghouse Owners Group participants in the Technical
Specification Task Force (TSTF) and is designated TSTF-370. TSTF-370 is
supported by NRC-approved topical report WCAP-15049-A, ``Risk-Informed
Evaluation of an Extension to Accumulator Completion Times,'' submitted
on May 18, 1999. The NRC staff issued a notice of opportunity for
comment in the Federal Register on July 15, 2002 (67 FR 46542), on
possible amendments concerning TSTF-370, including a model safety
evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on March 12, 2003 (68 FR 11880). The licensee affirmed the
applicability of the following NSHC determination in its application
dated July 14, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The basis for the accumulator limiting condition for operation
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of WCAP-15049-A, the proposed change will
allow plant operation with an inoperable accumulator for up to 24
hours, instead of 1 hour, before the plant would be required to
begin shutting down. The impact of the increase in the accumulator
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total
plant core damage frequency (CDF) less than 1.0E-03/yr. The
incremental conditional core damage probabilities calculated in
WCAP-15049-A for the accumulator CT increase meet the criterion of
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications,'' for all cases except those that are based on
design basis success criteria. As indicated in WCAP-15049-A, design
basis accumulator success criteria are not considered necessary to
mitigate large break loss-of-coolant accident (LOCA) events, and
were only included in the WCAP-15049-A evaluation as a worst case
data point. In addition, WCAP-15049-A states that the NRC has
indicated that an incremental conditional core damage frequency
(ICCDP) greater than 5E-07 does not necessarily mean the change is
unacceptable.
The proposed technical specification change does not involve any
hardware changes nor does it affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, engineered safety feature (ESF) actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049-A evaluation demonstrates that the small
increase in risk due to increasing the CT for an inoperable
accumulator is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed technical specification change. No new failure mode
has been created and no new equipment performance burdens are
imposed.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits.
The basis for the accumulator LCO, as discussed in Bases Section
3.5.1, is to ensure that a sufficient volume of borated water will
be immediately forced into the core through each of the cold legs in
the event the RCS pressure falls below the pressure of the
accumulators, thereby providing the initial cooling mechanism during
large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-
A, the proposed change will allow plant operation with an inoperable
accumulator for up to 24 hours, instead of 1 hour, before the plant
would be required to begin shutting down. The impact of this on
plant risk was evaluated and found to be very small. That is,
increasing the time the accumulators will be unavailable to respond
to a large LOCA event, assuming accumulators are needed to mitigate
the design basis event, has a very small impact on plant risk. Since
the frequency of a design basis large LOCA (a large LOCA with loss
of offsite power) would be significantly lower than the large LOCA
frequency of the WCAP-15049-A evaluation, the impact of increasing
the accumulator CT from 1 hour to 24 hours on plant risk due to a
design basis large LOCA would be significantly less than the plant
risk increase presented in the WCAP-15049-A evaluation.
Therefore, this change does not involve a significant reduction
in a margin of safety.
[[Page 59215]]
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: October 16, 2001; as supplemented by
letters dated May 20, September 12, and November 21, 2002; and January
27, and September 22, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to incorporate changes
resulting from the use of an alternate source term (AST) and the
implementation of several plant modifications. Publication of the
Proposed No Significant Hazards Consideration Determination and
Opportunity for Hearing for the October 16, 2001, submittal appeared in
the Federal Register on January 22, 2002, (67 FR 2922). The September
22, 2003, submittal contained a revised No Significant Hazards
Consideration Determination. The September 22, 2003, submittal includes
(1) Implementing the AST for accident analysis as described in
Regulatory Guide 1.183; (2) relaxing the TS for the penetration room
ventilation system (PRVS) and the spent fuel pool ventilation system
(SFPVS) because these systems are no longer credited for control room
and offsite doses; (3) revising the control room ventilation system
(CRVS) to allow for a one-time completion extension to support
implementation of the control room intake/booster fan modification; (4)
lowering the reactor building leakage rate from 0.25 weight percent per
day to 0.20 weight percent per day; (5) revising the ventilation filter
testing program radioactive methyl iodide removal acceptance criterion
for PRVS, SFPVS, and CRVS booster fan trains; and (6) adoption of TS
Task Force (TSTF)-51.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed amendment will not involve a significant
increase in the probability of consequences of an accident
previously evaluated.
The AST [alternate source term] and those plant systems affected
by implementing the proposed changes to the TS [technical
specifications] are not assumed to initiate design basis accidents.
The AST does not affect the design or operations of the facility.
Rather, the AST is used to evaluate the consequences of a postulated
accident. The implementation of the AST has been evaluated in the
revisions to the analysis of the design basis accident for ONS
[Oconee Nuclear Station]. Based on the results of these analyses, it
has been demonstrated that, with the requested changes, the dose
consequences of these events meet the acceptance criteria of 10 CFR
50.67 and RG [Regulatory Guide] 1.183. Therefore, the proposed
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
(2) The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The AST and those plant systems affected by implementing the
proposed changes to the TS are not assumed to initiate design basis
accidents. The systems affected by the changes are used to mitigate
the consequences of an accident that has already occurred. The
proposed TS changes and modifications do not significantly affect
the mitigative function of these systems. Consequently, these
systems do not alter the nature of events postulated in the Safety
Analysis Report nor do they introduce any unique precursor
mechanisms. Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) The proposed amendment will not involve a significant
reduction in the margin of safety.
The implementation of the AST, proposed changes to the TS and
implementation of the proposed modifications have been evaluated in
the revisions to the analysis of the consequences of the design
basis accidents for the ONS. Based on the results of these analyses,
it has been demonstrated that with the requested changes the dose
consequences of these events meet the acceptance criteria of 10 CFR
50.67 and following the provisions of RG 1.183. Thus, the proposed
amendment will not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois
Date of amendment request: July 29, 2003.
Description of amendment request: The proposed amendments would
allow the licensee to modify technical specifications (TS) to be
consistent with Technical Specification Task Force (TSTF) Traveler
TSTF-360, Revision 1, ``DC Electrical Rewrite,'' and to implement new
actions for inoperable battery chargers, modify certain actions and
surveillance requirements, relocate certain surveillance requirements
to a licensee controlled program, and create an administrative program
for battery monitoring and maintenance to be referenced in the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes restructure the Technical Specifications
(TS) for the direct current (DC) electrical power system. The
proposed changes add actions to specifically address battery charger
inoperability. This change will rely upon the capability of
providing the battery charger function by an alternate means (e.g.,
a 125 volts direct current (VDC) portable battery charger or a 250
VDC portable battery charger) to justify the proposed Completion
Times. The DC electrical power system, including associated battery
chargers, is not an initiator to any accident sequence analyzed in
the Updated Final Safety Analysis Report (UFSAR). Operation in
accordance with the proposed TS ensures that the DC electrical power
system is capable of performing its function as described in the
UFSAR. Therefore, the mitigative functions supported by the DC
electrical power system will continue to provide the protection
assumed by the analysis.
The relocation of preventive maintenance surveillance, and
certain operating limits and actions, to a newly-created licensee
controlled Battery Monitoring and Maintenance Program will not
challenge the ability of the DC electrical power system to perform
its design function. Appropriate monitoring and maintenance,
consistent with industry standards, will continue to be performed.
In addition, the DC electrical power system is within the scope of
10 CFR 50.65, ``Requirements for monitoring the effectiveness of
maintenance at nuclear power plants,'' which will ensure the control
of maintenance activities associated with the DC electrical power
system. The integrity of fission product barriers, plant
configuration,
[[Page 59216]]
and operating procedures as described in the UFSAR will not be
affected by the proposed changes. Therefore, the consequences of
previously analyzed accidents will not increase by implementing
these changes.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes involve restructuring the TS for the DC
electrical power system. This change will rely upon the capability
of providing the battery charger function by an alternate means
(e.g., a swing charger or a portable battery charger) to justify the
proposed Completion Times when a normal battery charger is
inoperable. The DC electrical power system, including associated
battery chargers, is not an initiator to any accident sequence
analyzed in the UFSAR. Rather, the DC electrical power system is
used to supply equipment used to mitigate an accident.
The 125 VDC portable battery charger will be utilized as a
common spare to feed the Division I or Division 2 125 VDC bus of
Unit 2 or Unit 3. For the 250 VDC system, a full capacity swing
charger is available for use between the units, and can be aligned
to any one of the 250 VDC batteries. In addition, the 250 VDC
portable battery charger can be utilized as a common spare to feed
the 250 VDC safety related batteries of Unit 2 or Unit 3. This
portable charger is identical to the existing chargers and is non-
safety related. The output of the portable charger will be capable
of being connected to any one of the Class IE DC buses for Division
I or Division 2 of Unit 2 or Unit 3. Allowing the use of a portable
spare and swing battery chargers will increase the reliability of
the DC electrical power system. The mitigative functions supported
by the DC electrical power system will continue to provide the
protection assumed by the safety analyses described in the UFSAR.
Therefore, there are no new types of failures that could be created
by a failure of the portable battery charger. As such, no new or
different kind of accident or transient is expected by these
changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed changes will not adversely affect
operation of plant equipment. These changes will not result in a
change to the setpoints at which protective actions are initiated.
Sufficient DC capacity to support operation of mitigation equipment
is ensured. The changes associated with the new Battery Maintenance
and Monitoring Program will ensure that the station batteries are
maintained in a highly reliable manner. The use of a portable
battery charger will increase the reliability of the DC system
during periods of normal battery charger inoperability. The
equipment fed by the DC electrical sources will continue to provide
adequate power to safety related loads in accordance with analysis
assumptions. Therefore, the proposed changes do not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Senior Counsel, Nuclear; Exelon Generation
Company LLC; 4300 Winfield Road; Warrenville, IL 60555.
NRC Section Chief: Anthony J. Mendiola.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: July 14, 2003.
Description of amendment request: The proposed change is requested
to support application of an alternative source term methodology, with
the exception that Technical Information Document 14844, ``Calculation
of Distance Factors for Power and Test Reactor Sites,'' will continue
to be used as the radiation dose basis for equipment qualification.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The implementation of alternative source term (AST) assumptions
has been evaluated in revisions to the analyses of the following
limiting design basis accidents (DBAs) at Peach Bottom Atomic Power
Station (PBAPS):
[sbull] Loss-of-Coolant Accident,
[sbull] Main Steam Line Break Accident,
[sbull] Fuel Handling Accident, and
[sbull] Control Rod Drop Accident.
Based upon the results of these analyses, it has been
demonstrated that, with the requested changes, the dose consequences
of these limiting events are within the regulatory guidance provided
by the NRC for use with the AST. This guidance is presented in 10
CFR 50.67 and associated Regulatory Guide 1.183, and Standard Review
Plan Section 15.0.1. The Alternative Source Term is an input to
calculations used to evaluate the consequences of an accident, and
does not by itself affect the plant response, or the actual pathway
of the radiation released from the fuel. It does however, better
represent the physical characteristics of the release, so that
appropriate mitigation techniques may be applied. Therefore, the
consequences of an accident previously evaluated are not
significantly increased.
The equipment affected by the proposed changes is mitigative in
nature, and relied upon after an accident has been initiated.
Application of the Alternative Source Term (AST) does not involve
any physical changes to the plant design. While the operation of
various systems do change as a result of these proposed changes,
these systems are not accident initiators. Application of the AST is
not an initiator of a design basis accident. The proposed changes to
the Technical Specifications (TS), while they revise certain
performance requirements, do not involve any physical modifications
to the plant. As a result, the proposed changes do not affect any of
the parameters or conditions that could contribute to the initiation
of any accidents. As such, removal of operability requirements
during the specified conditions will not significantly increase the
probability of occurrence for an accident previously analyzed. Since
design basis accident initiators are not being altered by adoption
of the Alternative Source Term analyses, the probability of an
accident previously evaluated is not affected.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed amendment does not involve a physical alteration of
the plant (no new or different type of equipment will be installed
and there are no physical modifications to existing equipment
associated with the proposed changes). Similarly, it does not
physically change any structures, systems or components involved in
the mitigation of any accidents, thus, no new initiators or
precursors of a new or different kind of accident are created. New
equipment or personnel failure modes that might initiate a new type
of accident are not created as a result of the proposed amendment.
As such the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Safety margins and analytical conservatisms have been evaluated
and have been found acceptable. The analyzed events have been
carefully selected and margin has been retained to ensure that the
analyses adequately bound postulated event scenarios. The dose
consequences due to design basis accidents comply with the
requirements of 10 CFR 50.67 and the guidance of Regulatory Guide
1.183.
The proposed amendment is associated with the implementation of
a new licensing basis for PBAPS Design Basis Accidents (DBAs).
Approval of the change from the
[[Page 59217]]
original source term to a new source term taken from Regulatory
Guide 1.183 is being requested. The results of the accident
analyses, revised in support of the proposed license amendment, are
subject to revised acceptance criteria. The analyses have been
performed using conservative methodologies, as specified in
Regulatory Guide 1.183. Safety margins have been evaluated and
analytical conservatism has been utilized to ensure that the
analyses adequately bound the postulated limiting event scenario.
The dose consequences of these DBAs remain within the acceptance
criteria presented in 10 CFR 50.67, ``Accident Source Term'', and
Regulatory Guide 1.183.
The proposed changes continue to ensure that the doses at the
exclusion area boundary (EAB) and low population zone boundary
(LPZ), as well as the Control Room, are within corresponding
regulatory limits.
Therefore, operation of PBAPS in accordance with the proposed
changes will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for Licensee: Mr. Edward Cullen, Vice President and
General Counsel, Exelon Generation Company, LLC, 2301 Market Street,
S23-1, Philadelphia, PA 19101.
NRC Section Chief: James W. Clifford.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 12, 2003.
Description of amendment request: The proposed amendment is for
relaxation of the heater acceptance criteria contained in Surveillance
Requirement (SR) 4.6.6.1d.5, SR 4.7.6.1d.3, and SR 4.7.7d.4 for the
shield building ventilation, control room ventilation, and controlled
ventilation area systems, respectively. These SRs are performed to
verify that heat dissipated by the heaters is within a given band. The
requested change is to increase the upper limit of the acceptance
criteria from rated capacity plus 5 percent (%) to rated capacity plus
10%. No change is proposed for the lower limit of the band of rated
capacity minus 10%.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The relaxation of the SR acceptance criteria to increase the
operating band does not alter the way plant equipment is designed or
operated. The ESF [engineered safety feature] filtration unit
heating coils will continue to reduce the humidity of the incoming
air to 70% relative humidity or below. In addition, the air
temperature will continue to be controlled such that additional
iodine will not be released into the environment. Thus, the charcoal
adsorber will continue to meet its design basis and its efficiency
will not be adversely affected. The effect of the higher heat
dissipation has also been evaluated and the ignition temperature of
the charcoal adsorbers is not approached with flow through the
systems. In addition, the impact of the new acceptance criterion was
determined not to impact the loading or fuel consumption of the
emergency diesel generators.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The relaxation of the SR acceptance criteria to increase the
operating band does not alter the way plant equipment is designed,
operated, or tested. No possibility for a new or different accident
or failure mode is introduced by modifying the SR acceptance
criteria. The proposed change does not affect the functional
capability of safety-related equipment.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The ESF filtration unit heating coils will continue to reduce
the humidity of the incoming air to 70% relative humidity or below.
Thus, the efficiency of the charcoal adsorber will not be adversely
affected. In addition, the impact of the new acceptance criterion
was determined not to impact the loading or fuel consumption of the
emergency diesel generators. Therefore, the systems have the same
capabilities to mitigate accidents as they had prior to the SR
acceptance criteria change.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: August 27, 2003.
Brief description of amendments: The proposed change allows entry
into a mode or other specified condition in the applicability of a
technical specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of 10 CFR 50.65(a)(4). Limiting
Condition for Operation (LCO) 3.0.4 exceptions in individual TS would
be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to
reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated August 27, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed
[[Page 59218]]
by proposed LCO 3.0.4, are no different than the consequences of an
accident while entering and relying on the required actions while
starting in a condition of applicability of the TS. Therefore, the
consequences of an accident previously evaluated are not
significantly affected by this change. The addition of a requirement
to assess and manage the risk introduced by this change will further
minimize possible concerns. Therefore, this change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: August 25, 2003.
Description of amendment request: The proposed license amendment
request would revise Technical Specification (TS) 3.5.1 to incorporate
TS Task Force 318 for one Low Pressure Coolant Injection (LPCI) pump
inoperable in each of the two Emergency Core Cooling Systems (ECCS)
divisions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not affect the LPCI subsystem design or
function. The change to TS 3.5.1 Condition A with one LPCI pump
inoperable in both subsystems is more reliable than the current
configuration allowed by Condition A. The current TS actions require
entry into shutdown LCO [Limiting Condition for Operation] 3.0.3 for
this condition. In addition, for an event that does not impact LPCI
availability the change provides for more injection flow than the
current TS 3.5.1 Condition A LPCI pump configuration. Review of
Updated Safety Analysis Report Section XIV-6.0 ``Analysis of Design
Basis Accidents'' confirms that the LPCI mode of the Residual Heat
Removal system is not assumed to be the initiator of any previously
analyzed event.
Based on the above, NPPD concludes that the proposed TS change
to TS 3.5.1 Condition A does not significantly increase the
probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical change to the
plant, add any new equipment or require any existing equipment to be
operated in a manner different from the present system design.
Based on the above, NPPD concludes that the proposed TS change
to TS 3.5.1 Condition A does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
The proposed TS change will not reduce the margin of safety. The
proposed configuration of one LPCI pump in each LPCI subsystem
represents a more reliable configuration. The current TS actions
require entry into shutdown LCO 3.0.3 for this condition. In
addition, for an event that does not impact LPCI availability the
change provides for more injection flow than the current [LCO]
requirement which only allows two LPCI pumps in one ECCS subsystem
to be inoperable for seven days.
Based on the above, NPPD concludes that the proposed TS change
to TS 3.5.1 Condition A does not involve a significant reduction in
the margin of safety.
From the above discussions, NPPD concludes that the proposed
amendment involves no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: September 18, 2003.
Description of amendment request: The proposed amendment would
revise the limiting condition for operation (LCO) and the associated
surveillance requirements of Technical Specification 3.4.1, ``[Primary
Coolant System] PCS Pressure, Temperature, and Flow Departure from
Nucleate Boiling (DNB) Limits,'' to reflect relocation of the DNB
limits from the TSs to the Core Operating Limits Report (COLR). These
DNB limits are for pressurizer pressure, PCS cold leg temperature, and
PCS total flow rate. The proposed amendment would also revise paragraph
a of TS 5.6.5, ``Core Operating Limits Report (COLR),'' to reflect the
addition of ``DNB Limits'' to the COLR. In addition, LCO 3.4.1 would be
added to items 16 and 17 in TS 5.6.5b, which lists the documents
approved by the NRC for the analytical methods for which the licensee
is to use the latest revisions to determine the core operating limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 59219]]
Response: No
The proposed amendment relocates the primary coolant system
(PCS) departure from nucleate boiling (DNB) limits to the core
operating limits report (COLR) and does not involve any change to
the PCS DNB limits themselves. The proposed amendment does not
involve operation of any required structures, systems, or components
(SSCs) in a manner or configuration different from those previously
recognized or evaluated. The Nuclear Regulatory Commission (NRC) has
approved all the analytical methods described in Technical
Specification (TS) section 5.6.5, ``Core Operating Limits Report
(COLR).'' Relocation of the PCS DNB limits to the COLR will maintain
existing operating fuel cycle analysis requirements. Any future
revisions to the safety analyses that require prior NRC approval are
identified per the 10 CFR 50.59 review process.
Therefore, the probability of an accident previously evaluated
will not be increased by the proposed change.
The consequences of an accident previously evaluated will not be
increased since the reactor is still protected from violating the
PCS DNB parameters used in the safety analysis for Palisades Nuclear
Plant.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment to relocate the PCS DNB limits to the
COLR would not change or add a system function. The proposed
amendment does not involve operation of any required SSCs in a
manner or configuration different from those previously recognized
or evaluated. No new failure mechanisms will be introduced by the
proposed change.
Therefore, this proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment to relocate the PCS DNB limits to the
COLR will continue to assure that the acceptance criteria
established in the safety analysis will be met. The safety analyses
of normal operating conditions and anticipated operational
occurrences assume initial conditions within the normal steady state
envelope. The limits placed on DNB related parameters ensure that
these parameters, when appropriate measurement uncertainties are
applied, will not be less conservative than those assumed in the
safety analyses and thereby provide assurance that the minimum
departure from nucleate boiling ratio (DNBR) will meet the required
criteria for each of the analyzed transients. The proposed amendment
does not change the existing PCS DNB limits. Any future revisions to
the safety analyses that require prior NRC approval are identified
per the 10 CFR 50.59 review process.
Therefore, the proposed amendment would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: L. Raghavan.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: September 15, 2003.
Description of amendment requests: In Technical Specification (TS)
2.0, ``Safety Limits (SLs),'' Reactor Core SL 2.1.1.2, the proposed
change would replace the peak linear heat rate SL with a peak fuel
centerline temperature SL. This change is requested so SL 2.1.1.2
adequately conforms to 10 CFR 50.36(c)(1)(ii)(A), which requires that
Limiting Safety System Settings prevent a Safety Limit from being
exceeded.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not require any physical change to any
plant systems, structures, or components nor does it require any
change in systems or plant operations. The proposed change does not
require any change in safety analysis methods or results. The change
to establish the PFCT [Peak Fuel Centerline Temperature] as the SL
is consistent with the Standard Review Plan (SRP) and the SONGS
Units 2 and 3 licensing basis for ensuring that the fuel design
limits are met. Operations and analysis will continue to be in
compliance with NRC regulations.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The SONGS Units 2 and 3 Updated Final Safety Analysis Report
(UFSAR) Chapter 15 accident analysis for Anticipated Operational
Occurrences (AOOs) where the peak linear heat rate may exceed the
existing Safety Limit of 21 KW/ft is the Control Element Assembly
(CEA) Withdrawal at subcritical and low power startup conditions.
The accident analyses indicate that the peak linear heat rate
may exceed the Limiting Safety System Setpoint of 21 KW/ft during
Control Element Assembly Withdrawal Events at Subcritical and Hot
Zero Power conditions. The analyses for these AOOs indicate that the
PFCT is not approached or exceeded. The existing analyses remain
unchanged and do not affect any accident initiators that would
create a new accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not require any change in accident
analysis methods or results. Therefore, by changing the SL from PLHR
[Peak Linear Heat Rate] to Peak Fuel Centerline Temperature, the
margin as established in the current license basis remains
unchanged.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station
(VCSNS), Unit No. 1, Fairfield County, South Carolina
Date of amendment request: July 29, 2003.
Description of amendment request: The proposed change will revise
Surveillance Requirement 4.0.5 to reflect the deletion of Subsections
IWP and IWV from Section XI of the 2000 Addenda of American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. This
change will also result in revising the Technical Specification (TS)
Bases for 4.0.5, 3/4.4.2 and 3/4.4.6 to reflect the applicability of
the Code for Operation and Maintenance of Nuclear Power Plants (OM
Code) to inservice testing activities. TS 4.0.5 is also being revised
as recommended by NUREG-1492, ``Guidelines for Inservice Testing at
Nuclear Power Plants,'' April 1995.
[[Page 59220]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to TS 4.0.5 reflects NRC approval of the
ASME Code [2000 Adenda], in 10CFR50.55a, for the conduct of
Inservice Testing (IST). The current TS references use of ASME
Section XI for this testing, which will no longer be applicable for
the third IST interval. The adoption of an NRC approved test code,
as required by 10CFR50.55a(f)(4)(ii) will not increase the
probability of an accident previously evaluated. Testing is
performed to ensure the operational readiness of pumps and valves to
perform their safety functions.
The probability or consequences of accidents previously
evaluated in the VCSNS FSAR [Final Safety Analysis Report] are
unaffected by this proposed change because there is no change to any
equipment response or accident mitigation scenario. There are no
additional challenges to fission product barrier integrity.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change involves the adoption of an NRC approved
Inservice Testing Code for the conduct of Operating License mandated
testing. The adoption of the new Code is required to satisfy
10CFR50.55a(f)(4)(ii). The new Code enhances plant safety by
requiring the bi-directional testing of check valves and
comprehensive pump testing. These changes were incorporated to
better monitor pumps and check valves for degradation. The adoption
of the new Code does not create the possibility of a new or
different kind of accident or malfunction.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
The proposed change does not challenge the performance or integrity
of any safety-related system. Therefore, the proposed change does
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Does this change involve a significant reduction in margin of
safety?
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change will have no affect
on the availability, operability, or performance of the safety-
related systems and components. A change to the surveillance
requirement is proposed, but the ASME OM Code is an NRC approved
standard incorporating inservice testing enhancements not contained
in ASME Section XI.
Pursuant to 10 CFR 50.91, the preceding analyses provide a
determination that the proposed Technical Specifications change
poses no significant hazard as delineated by 10 CFR 50.92.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas G. Eppink, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: John A. Nakoski.
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: September 2, 2003.
Description of amendment request: The proposed change involves the
extension from 1 hour to 24 hours of the completion time (CT) for
Condition B of Technical Specification (TS) 3.5.1, which defines
requirements for accumulators. Accumulators are part of the emergency
core cooling system and consist of tanks partially filled with borated
water and pressurized with nitrogen gas. The contents of the tank are
discharged to the reactor coolant system if, as during a loss-of-
coolant accident, the coolant pressure decreases to below the
accumulator pressure. Condition B of TS 3.5.1 specifies a CT to restore
an accumulator to operable status when it has been declared inoperable
for a reason other than the boron concentration of the water in the
accumulator not being within the required range. This change was
proposed by the Westinghouse Owners Group participants in the Technical
Specification Task Force (TSTF) and is designated TSTF-370. TSTF-370 is
supported by NRC-approved topical report WCAP-15049-A, ``Risk-Informed
Evaluation of an Extension to Accumulator Completion Times,'' submitted
on May 18, 1999. The NRC staff issued a notice of opportunity for
comment in the Federal Register on July 15, 2002 (67 FR 46542), on
possible amendments concerning TSTF-370, including a model safety
evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on March 12, 2003 (68 FR 11880). The licensee affirmed the
applicability of the following NSHC determination in its application
dated September 2, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The basis for the accumulator limiting condition for operation
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of WCAP-15049-A, the proposed change will
allow plant operation with an inoperable accumulator for up to 24
hours, instead of 1 hour, before the plant would be required to
begin shutting down. The impact of the increase in the accumulator
CT on core damage frequency for all the cases evaluated in WCAP-
15049-A is within the acceptance limit of 1.0E-06/yr for a total
plant core damage frequency (CDF) less than 1.0E-03/yr. The
incremental conditional core damage probabilities calculated in
WCAP-15049-A for the accumulator CT increase meet the criterion of
5E-07 in Regulatory Guides (RG) 1.174, ``An Approach for using
Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-
Specific Changes to the Licensing Basis,'' and 1.177, ``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications,'' for all cases except those that are based on
design basis success criteria. As indicated in WCAP-15049-A, design
basis accumulator success criteria are not considered necessary to
mitigate large break loss-of-coolant accident (LOCA) events, and
were only included in the WCAP-15049-A evaluation as a worst case
data point. In addition, WCAP-15049-A states that the NRC has
indicated that an incremental conditional core damage frequency
(ICCDP) greater than 5E-07 does not necessarily mean the change is
unacceptable.
The proposed technical specification change does not involve any
hardware changes nor does it affect the probability of any event
initiators. There will be no change to normal plant operating
parameters, engineered safety feature (ESF) actuation setpoints,
accident mitigation capabilities, accident analysis assumptions or
inputs.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
[[Page 59221]]
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049-A evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049-A evaluation demonstrates that the small
increase in risk due to increasing the CT for an inoperable
accumulator is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed technical specification change. No new failure mode
has been created and no new equipment performance burdens are
imposed.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits.
The basis for the accumulator LCO, as discussed in Bases Section
3.5.1, is to ensure that a sufficient volume of borated water will
be immediately forced into the core through each of the cold legs in
the event the RCS pressure falls below the pressure of the
accumulators, thereby providing the initial cooling mechanism during
large RCS pipe ruptures. As described in Section 9.2 of WCAP-15049-
A, the proposed change will allow plant operation with an inoperable
accumulator for up to 24 hours, instead of 1 hour, before the plant
would be required to begin shutting down. The impact of this on
plant risk was evaluated and found to be very small. That is,
increasing the time the accumulators will be unavailable to respond
to a large LOCA event, assuming accumulators are needed to mitigate
the design basis event, has a very small impact on plant risk. Since
the frequency of a design basis large LOCA (a large LOCA with loss
of offsite power) would be significantly lower than the large LOCA
frequency of the WCAP-15049-A evaluation, the impact of increasing
the accumulator CT from 1 hour to 24 hours on plant risk due to a
design basis large LOCA would be significantly less than the plant
risk increase presented in the WCAP-15049-A evaluation.
Therefore, this change does not involve a significant reduction in
a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorneys for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201; Mr. Arthur H. Domby, Troutman Sanders, NationsBank
Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, Georgia 30308-
2216.
NRC Section Chief: John A. Nakoski.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 22, 2003.
Description of amendment request: The proposed amendment revises
Technical Specification 3.3.2 governing radiation monitoring
instrumentation to relax restrictions on containment purge valve
operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The radiation monitors affected by the proposed amendment are
not potential accident initiators. Adequate measures are available
to compensate for instrumentation that is out of service. The
proposed amendment does not affect how the affected instrumentation
normally functions or its role in the response of an operator to an
accident or transient. The core damage frequency in the STP [South
Texas Project] PRA [probabilistic risk assessment] is not impacted
by the proposed changes. Therefore, STPNOC [South Texas Project
Nuclear Operating Company] concludes that there is no significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The instrumentation affected by the proposed amendment is not
credited for the prevention of any accident not evaluated in the
safety analysis. The proposed amendment involves no changes in the
way the plant is operated or controlled. It involves no change in
the design configuration of the plant. No new operating environments
are created. Therefore, STPNOC concludes the proposed change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change has no significant effect on functions that
are supported by the affected instrumentation. There will be no
significant effect on the availability and reliability of the
affected instrumentation. Adequate measures are available to
compensate for instrumentation that is out of service. Therefore,
STPNOC concludes the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: August 7, 2003.
Brief description of amendments: The proposed change allows entry
into a mode or other specified condition in the applicability of a
technical specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of 10 CFR 50.65(a)(4). Limiting
Condition for Operation (LCO) 3.0.4 exceptions in individual TS would
be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to
reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated August 7, 2003.
[[Page 59222]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: July 18, 2003.
Brief description of amendments: The proposed change allows entry
into a mode or other specified condition in the applicability of a
technical specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of 10 CFR 50.65(a)(4). Limiting
Condition for Operation (LCO) 3.0.4 exceptions in individual TS would
be eliminated, and Surveillance Requirement 3.0.4 revised to reflect
the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated July 18, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS Limiting
Conditions for Operation (LCO). The risk associated with this
allowance is managed by the imposition of required actions that must
be performed within the prescribed completion times. The net effect
of being in a TS condition on the margin of safety is not considered
significant. The proposed change does not alter the required actions
or completion times of the
[[Page 59223]]
TS. The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: August 29, 2003.
Brief description of amendment request: The proposed amendment
would revise the Updated Final Safety Analysis Report to use the
reactor building crane for heavy loads up to a total of 117 tons for
removal and reinstallation activities for the reactor shield blocks
prior to and during the Units 2 outage D2R18.
Date of publication of individual notice in Federal Register:
September 10, 2003.
Expiration date of Individual notice: October 10, 2003.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: April 2, 2001, as supplemented
by letters dated January 15, and August 23, 2002, March 28, and August
19, 2003.
Brief description of amendment: The amendment identifies the
conditions under which the inclined fuel transfer system blind flange
may be removed when primary containment integrity is required (i.e.,
during Modes 1, 2, and 3) and restricts this configuration to no more
than 40 days per operating cycle. These changes are reflected by (1)
adding Note 5 for the Actions of Technical Specification (TS) 3.6.1.3,
``Primary Containment Isolation Valves (PCIVs),'' (2) deleting Note 3
of TS Surveillance Requirement 3.6.1.3.3, (3) adding a conditional note
to TS 3.6.1.1, ``Primary Containment--Operating,'' and (4) associated
TS Bases changes.
Date of issuance: September 17, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 158.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 13, 2003 (68 FR
25650). The supplemental letter of August 19, 2003, contained
clarifying information and did not change the initial no significant
hazards consideration determination and did not expand the scope of the
original Federal Register Notice. The Commission's related evaluation
of the amendment is contained in a Safety Evaluation dated September
17, 2003.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: July 31, 2002, and supplemented
by letters dated March 7 and August 28, 2003.
Brief description of amendment: The amendment revises Appendix A,
Technical Specifications (TSs), of the Operating License by adding a
Surveillance Requirement (SR) to TS 3.2.2, ``Minimum Critical Power
Ratio (MCPR),'' that requires determination of the MCPR limits
following completion of control rod scram time testing. The new SR
provides for the required evaluation necessary to apply faster scram
times to provide for improved MCPR operating limits.
Date of issuance: September 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 159.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58637). The supplemental letters
[[Page 59224]]
contained clarifying information and did not change the initial no
significant hazards consideration determination and did not expand the
scope of the original Federal Register Notice. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
September 29, 2003.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: January 14, 2003.
Brief description of amendment: The amendment revised technical
specification sections 3.8.9, 3.15.2, 4.12.2, and associated Bases to
delete the requirements for the reactor building purge air treatment
system.
Date of issuance: September 23, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 245.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: (68 FR 10278) March 4,
2003. The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 23, 2003.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: September 20, 2002.
Brief description of amendment: The amendment revises the Technical
Specification (TS) definition of containment integrity to ensure that
all power-operated valves, relief valves, and check valves are included
and clarifies the handling of operability and reportability issues
related to Type III containment isolation valves. The amendment also
includes minor administrative and editorial changes to improve the
consistency and clarity of the TSs.
Date of issuance: September 30, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 246.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68729). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 30, 2003.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit No. 2, Maricopa County, Arizona
Date of application for amendment: December 21, 2001, as
supplemented by letters dated March 13, August 27, August 29, September
4, September 6, October 11, November 21, December 10, December 23,
2002, and March 11, June 10, July 25, and August 22, 2003.
Brief description of amendment: The amendment changes the Unit 2
Technical Specifications and operating license to support (1)
replacement of the steam generators and (2) the subsequent operation at
an increased maximum power level of 3990 MWt, which is a 2.94 percent
increase from the current 3876 MWt.
Date of issuance: September 29, 2003.
Effective date: This license amendment is effective as of the date
of issuance, and shall be implemented prior to entry into Mode 4 during
the restart from the Fall 2003 refueling outage.
Amendment No.: Unit 2-149.
Facility Operating License No. NPF-51: The amendment revised the
Technical Specifications and Operating License.
Date of initial notice in Federal Register: February 19, 2002 (67
FR 7412). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 29, 2003.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: August 14, 2002, as supplemented
on March 11, May 16, and May 23, 2003.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) related to reactivity control systems, power
distribution limits, and special test exceptions.
Date of issuance: September 25, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 280.
Facility Operating License No. DPR-65: This amendment revised the
TSs.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58640). The supplements dated March 11, May 16, and May 23, 2003,
provided additional information which clarified the application, did
not expand the scope of the application as originally noticed, and did
not change the staff's original proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated September 30, 2003.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: August 7, 2002, as supplemented
on October 23, 2002.
Brief description of amendment: The amendment revises Technical
Specification (TS) 6.9.1.8, ``Core Operating Limits Report,'' to update
the list of documents that describe the analytical methods used to
determine the core operating limits.
Date of issuance: September 25, 2003.
Effective date: As of the date of issuance and shall be implemented
prior to Mode 4 operation of Cycle 16.
Amendment No.: 281.
Facility Operating License No. DPR-65: This amendment revised the
TSs.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58639). The supplement dated October 23, 2002, provided additional
information which clarified the application, did not expand the scope
of the application as originally noticed, and did not change the
staff's original proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 25, 2003.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: May 7, 2002, as supplemented on
January 16, May 27, July 1, and August 21, 2003.
Brief description of amendment: The amendment revises Technical
Specifications (TSs) 2.2, ``Limiting Safety System Settings'' and 3/
4.3, ``Instrumentation'' to more accurately reflect the existing plant
design for the Reactor Protection System, the Engineered Safety
Features Actuation System, and the Radiation Monitoring System
instrumentation and to provide
[[Page 59225]]
consistency within the associated TS Tables.
Date of issuance: September 25, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 282.
Facility Operating License No. DPR-65: This amendment revises the
TSs.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42819). The supplements dated January 16, May 27, July 1, and August
21, 2003, provided additional information which clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated September 25, 2003.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: August 14, 2002, as supplemented
December 19, 2002.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) related to Containment Systems. Specifically, the
revisions: (1) Added clarification to TS 1.7, ``Definitions--
Containment Integrity;'' (2) added clarifying information, as well as
revised a portion of Surveillance Requirement 4.6.1.1 associated with
the affected section of TS 3.6.1.1, ``Containment Integrity;'' (3)
revised TS 3.6.3, ``Containment Isolation Valves,'' that made editorial
changes, added clarifying information, and added an Action item that
increased the allowed outage time from 4 hours to 72 hours for
Containment Isolation Valves in closed systems; and (4) made other
changes that were clarifying and/or administrative in nature. In
addition, the TS Bases were revised to address these changes, as
appropriate.
Date of issuance: September 29, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 90 days from the date of issuance.
Amendment No.: 216.
Facility Operating License No. NPF-49: This amendment revised the
TSs.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61678). The December 19, 2002, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the amendment beyond the scope of
the initial notice. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated September 29, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: December 20, 2001, as
supplemented by letters dated March 4, 2002, September 12, 2002,
November 20, 2002, and August 28, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) 3.3.2, Engineered Safety Features
Actuation System Instrumentation.
Date of issuance: September 10, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 208 and 202.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 19, 2002 (67 FR
12601). The supplements dated March 4, 2002, September 12, 2002,
November 20, 2002, and August 28, 2003, provided clarifying information
that did not change the scope of the December 20, 2001, application or
the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 10, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: January 31, 2003, as
supplemented by letters dated June 12, and September 2, 2003.
Brief description of amendments: The amendments revise the
Technical Specifications to incorporate revised means of determining
the mass of ice in the ice condenser containment.
Date of issuance: September 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 209 and 203.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003, (68 FR
18274). The supplements dated June 12, and September 2, 2003, provided
clarifying information that did not change the scope of the January 31,
2003, application nor the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated September 29,
2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: January 31, 2003, as
supplemented by letters dated June 12, and September 2, 2003.
Brief description of amendments: The amendments revise the
Technical Specifications to incorporate revised means of determining
the mass of ice in the ice condenser containment.
Date of issuance: September 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 217 and 199.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003, (68 FR
18274). The supplements dated June 12, and September 2, 2003, provided
clarifying information that did not change the scope of the January 31,
2003, application nor the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated September 29,
2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: March 20, 2003, supplemented by
letters dated July 22, and August 5, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications and the licensing basis in the Updated Safety
Analysis Report to support installation of a passive low-pressure
injection cross connect inside containment.
Date of Issuance: September 29, 2003.
Effective date: As of the date of issuance and shall be implemented
[[Page 59226]]
within 90 days from the date of issuance.
Amendment Nos.: 335, 335, and 336.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
22745). The supplement dated July 22 and August 5, 2003, provided
clarifying information that did not change the scope of the March 20,
2003, application nor the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated September 29,
2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: August 16, 2002, as supplemented
June 6, 2003.
Brief description of amendment: The amendment adds a new Technical
Specification (TS) requirement to the Pilgrim Nuclear Power Station
(Pilgrim) TSs consistent with Technical Specification Task Force
(TSTF)-358. TSTF-358 addresses modifications to requirements for missed
surveillances consistent with NUREG 1433, Revision 2, ``Standard
Technical Specification, General Electric Plants, BWR/4'' (STS)
surveillance requirement 3.0.3. The amendment to the Pilgrim TSs is
added as TS 4.0.3.
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of opportunity for comment in the Federal Register on June 14, 2001 (66
FR 32400), on possible amendments concerning missed surveillances,
including a model safety evaluation (SE) and model no significant
hazards consideration (NSHC) determination, using the consolidated line
item improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on September 28, 2001 (66 FR
49714). The licensee affirmed the applicability of the model NSHC
determination in its application dated August 16, 2002, as supplemented
on June 6, 2003.
In addition, the following statement was added to the TS definition
of Limiting Condition for Operation (LCO): ``Failure to meet a
Surveillance, whether such failure is experienced during the
performance of the Surveillance or between performances of the
Surveillance, shall be failure to meet the LCO.'' The amendment also
made administrative changes to add new TS Sections 3.0, ``Limiting
Condition for Operation (LCO) Applicability,'' and 4.0, ``Surveillance
Requirement (SR) Applicability,'' into the Pilgrim TSs. New TSs 3.0,
4.0.1, and 4.0.2 are identified as ``Not Used.'' These changes rectify
the differences in the format and terminology of the current Pilgrim
TSs compared to the STS. The associated Bases are also implemented.
Date of issuance: September 30, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 203.
Facility Operating License No. DPR-35: The amendment revised the
TSs.
Date of initial notice in Federal Register: July 22, 2003 (68 FR
43390). The Commission's related evaluation of the amendment is
contained in a SE dated September 30, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 11, 2003.
Brief description of amendment: The amendment revises and relocates
Surveillance Requirement (SR) 4.0.5 and SR 4.4.9 to the administrative
section of the Technical Specifications (TS) under sections 6.5.8 and
6.5.7, respectively. The amendment also relocates TS 3.4.9, ``Reactor
Coolant System Structural Integrity'' and its Bases to the Technical
Requirements Manual. Additionally, the amendment extends the Waterford
3 flywheel volumetric examination interval to ten years.
Date of issuance: September 22, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 189.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 27, 2003 (68 FR
28851). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 22, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of application for amendments: April 19, 2002, and as
supplemented September 9, 2002, January 3, and July 13, 2003.
Brief description of amendments: The amendments would revise the
surveillance frequency of the containment spray system nozzles from 10
years to ``Following maintenance that could result in nozzle blockage,
OR Following fluid flow through the nozzles.''
Date of issuance: September 22, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 134 and 134.
Facility Operating License Nos. NPF-37 and NPF-66: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 11, 2002 (68 FR
40023). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 22, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: October 28, 2002.
Brief description of amendments: The amendments authorize changes
to the Updated Final Safety Analysis Report to describe the use of cast
iron materials in the containment cooling service water and diesel
generator cooling water systems.
Date of issuance: September 17, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 201 and 193.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revise the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75875). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 17, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: December 4, 2001.
Brief description of amendment: The amendment revises the Davis-
Besse Nuclear Power Station Operating License, Appendix A, Technical
Specifications (TS) Section 6.9, ``Administrative Controls--Reporting
Requirements,'' to eliminate the requirement to submit startup test
reports to the NRC.
[[Page 59227]]
Date of issuance: September 25, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 258.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 22, 2003 (68 FR
43391). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 25, 2003.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: July 18, 2002.
Brief description of amendments: These amendments revise the
Technical Specifications regarding the time period that inoperable
channels of the engineered safety feature actuation system can be in
the bypassed or tripped condition.
Date of Issuance: September 30, 2003.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 188 and 132.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: August 20, 2002 (67 FR
53987). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 30, 2003.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: October 11, 2002, as supplemented by
letters dated April 21, and July 29, 2003.
Description of amendment request: The amendment revises the
Technical Specifications (TS) to eliminate the Power Range Neutron Flux
High Negative Rate Reactor Trip function from TS 3/4.3.1, ``Reactor
Trip System Instrumentation,'' TS 2.2.1, ``Reactor Trip System
Instrumentation Setpoints,'' and their associated Bases. The amendment
also revises TS 3/4.10.3, ``Physics Tests,'' TS 3/4.10.4, ``Reactor
Coolant Loops,'' and TS Table 4.3-1, ``Reactor Trip System
Instrumentation Surveillance Requirements,'' that are associated with
certain testing activities required during STARTUP operations. The
revision also rewords the time interval for the Analog Channel
Operational Test (ACOT) in surveillance requirement (SR) 4.10.3.2. In
correlation with the revision to extend the ACOT interval in SR
4.10.3.2, Table 4.3-1 Note 1 is revised. This revision also extends the
ACOT interval for those Functional Units that reference TS Table 4.3-1
Note 1. The revision to TS 3/4.10.4 will delete TS 3/4.10.4 in its
entirety. Additionally, as a result of deleting TS 3/4.10.4, the
footnote which references TS 3/4.10.4 in TS 3/4.4.1.1 is deleted as
well.
Date of issuance: October 1, 2003.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 91.
Facility Operating License No. NPF-86: Amendment revises the TSs.
Date of initial notice in Federal Register: November 26, 2002 (67
FR 70767). The April 21, 2003 and July 16, 2003, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the amendment
beyond the scope of the initial notice. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
October 1, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: September 30, 2002, as
supplemented July 24 and September 25, 2003.
Brief description of amendment: The amendment authorizes changes to
the Updated Safety Analysis Report (USAR) to allow the use of an
upgraded computer code for design-basis accident containment integrity
analyses called Generation of Thermal-Hydraulic Information for
Containment (GOTHIC) version 7.0p2 (GOTHIC 7) with noted conditions.
Date of issuance: September 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 169.
Facility Operating License No. DPR-43: The amendment authorizes
changes to the Updated Safety Analysis Report.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66011). The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated September 29, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: April 30, 2003.
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant, Technical Specification Section 6.3, ``Plant Staff
Qualifications.'' The amendment updates requirements that have been
outdated based on licensed operator training programs being accredited
by the National Academy for Nuclear Training and promulgation of the
revised 10 CFR 55, ``Operators' Licenses.''
Date of issuance: October 2, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 170.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34670). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 2, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: March 11, 2003, as supplemented
July 16, 2003.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.1.4, ``Rod Group Alignment Limits,'' and TS 3.1.7,
``Rod Position Indication,'' to add a 1-hour soak time to both TSs to
allow the control rod drive mechanisms additional time following
substantial rod motion to reach thermal equilibrium.
Date of issuance: October 1, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 160 and 151.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the TSs.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18280). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 1, 2003.
[[Page 59228]]
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County,
California
Date of application for amendments: April 15, 2002, as supplemented
by letters dated September 27, 2002, February 28, 2003, April 25, 2003,
June 24, 2003, and September 12, 2003.
Brief description of amendments: The amendments authorize changes
to the Final Safety Analysis Report (FSAR) Update, together with other
analyses, design, and procedure changes, to implement the Diablo Canyon
Power Plant NUREG-0612, ``Control of Heavy Loads at Nuclear Power
Plants'' program that is required to implement a dry cask Independent
Spent Fuel Storage Installation (ISFSI).
Date of issuance: September 26, 2003.
Effective date: September 26, 2003, and shall be implemented
following the implementation of the ISFSI. The implementation of the
amendments include the incorporation into the FSAR Update the changes
discussed above, as described in the licensee's application dated April
15, 2002; its supplements dated September 27, 2002, February 28, 2003,
April 25, 2003, June 24, 2003, and September 12, 2003; and evaluated in
the staff's safety evaluation attached to the amendment.
Amendment Nos.: 162 and 163.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
authorized revision of the FSAR Update.
Date of initial notice in Federal Register: June 11, 2002 (67 FR
40025) The supplemental letters dated September 27, 2002, February 28,
2003, April 25, 2003, June 24, 2003, and September 12, 2003, provided
additional clarifying information, did not expand the scope of the
application as originally noticed, and did not change the NRC staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 26, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: April 10, 2003.
Brief description of amendments: The amendments revise Salem
Nuclear Generating Station, Unit Nos. 1 and 2 (Salem), Technical
Specifications (TSs) Table 3.3-1, ``Reactor Trip System
Instrumentation,'' by modifying the ``Condition and Setpoint''
description of permissive interlock ``P-7.'' The phrase ``Turbine
impulse chamber pressure,'' contained in the ``Condition and Setpoint''
description for permissive P-7, is replaced with the phrase ``Turbine
steam line inlet pressure'' in order to support planned modifications
to Salem's high pressure turbines.
Date of issuance: October 1, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 259 and 240.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34672). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 1, 2003.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: September 24, 2002, supplemented
by letters dated April 8 and May 21, 2003.
Brief description of amendment: This amendment revises the Action
Statement and surveillance requirements for the emergency diesel
generators (EDGs). The proposed changes would revise TS Section
3.8.1.1, Action b.2 and Action c.2, and TS Section 4.8.1.1, `` AC
Sources'' and associated Bases Section related to the EDG.
Date of issuance: September 26, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 164.
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68742). The April 8 and May 21, 2003, letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the scope of the
application. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 26, 2003.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: December 19, 2002, as
supplemented by letters dated April 7, May 21, May 30, June 4,
September 4, and September 12, 2003.
Brief description of amendments: The amendments revise the licensed
power level for Hatch, Units 1 and 2 by 1.5 percent from 2763 megawatts
thermal (MWt) to 2804 MWt. The change is based on the installation of
the Advanced Measurement Analysis Group, Inc. (AMAG)/Westinghouse
Crossflow ultrasonic flow measurement instrumentation, resulting in
improved feedwater flow measurement accuracy. The amendment changes the
Renewed Facility Operating License (RFOL) and the Technical
Specifications (TSs) to reflect the increased licensed power level.
Date of issuance: September 23, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 238 and 180.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revise the RFOL and the TSs.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7821). The supplements dated April 7, May 21, May 30, June 4,
September 4, and September 12, 2003, provided clarifying information
that did not change the scope of the December 19, 2002, application nor
the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 23, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: February 14, 2003, as
supplemented by letters dated June 5 and August 21, 2003.
Brief description of amendment: The amendment consists of changes
to Technical Specification (TS) 5.9.5, ``Core Operating Limits Report
(COLR).'' The revised TS modifies TS 5.9.5 to add three additional
methodologies in support of the Westinghouse 17x17
[[Page 59229]]
Robust Fuel Assembly (RFA)-2 fuel design with Intermediate Flow Mixers.
Date of issuance: September 30, 2003.
Effective date: As of the date of issuance and shall be implemented
no later than MODE 6 entry following the next refueling outage in the
fall of 2003.
Amendment No.: 46.
Facility Operating License No. NPF-90: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 1, 2003 (68 FR
15765). The supplemental letters provided clarifying information that
did not expand the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated September 30, 2003.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: March 6, 2003, as supplemented by
letters dated July 25, August 29, and September 16, 2003.
Brief description of amendments: The amendments revise the Final
Safety Analysis Report (FSAR) and Technical Specification (TS) Bases
reflecting approval of elimination of response time testing for
selected Reactor Trip System and Engineered Safety Features Actuation
System protection channel equipment.
Date of issuance: September 25, 2003.
Effective date: As of the date of issuance. The TS Bases shall be
implemented within 60 days from the date of issuance and the FSAR shall
be implemented in the next periodic update to the FSAR in accordance
with 10 CFR 50.71(e).
Amendment Nos.: 107 and 107.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the FSAR and TS Bases.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18288). The July 25, August 29, and September 16, 2003, supplemental
letters provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register on
April 15, 2003 (68 FR 18288). The Commission's related evaluation of
the amendments is contained in a Safety Evaluation dated September 25,
2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 3rd day of October.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-25742 Filed 10-10-03; 8:45 am]
BILLING CODE 7590-01-P